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Journal Articles

Investigation on natural circulation behavior for decay heat removal in reactor vessel of sodium-cooled fast reactor under severe accident condition, 2; Transient behavior under operations of multiple decay heat removal systems

Aizawa, Kosuke; Tsuji, Mitsuyo; Kobayashi, Jun; Kurihara, Akikazu

Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 7 Pages, 2022/10

In sodium-cooled fast reactors (SFRs), optimizing the design and operate decay heat removal systems (DHRSs) is important for safety enhancement against severe accidents. Thus, it is required to evaluate the cooling capability of DHRSs including the natural circulation behavior inside the reactor vessel during heat-removal phase that the fuel debris relocated in the reactor vessel is cooled by DHRSs. In this study, the experiments which simultaneously operations of the dipped-type DHX and the penetrated-type DHX were conducted to investigate the effect of operating multiple decay heat removal system on the natural circulation behavior in the reactor vessel. After achieving the stable conditions by operating the dipped-type DHX or the penetrated-type DHX, the other DHX was operated and the transient behavior was clarified by the temperature measurements. The clear temperature rise in the reactor vessel was confirmed by operating the penetrated-type DHX as second DHX operation under the condition of the dipped-type DHX operation at the beginning and the high heater power of fuel debris on the core catcher. Therefore, it was confirmed that the inhibition of the cooling for the decay heat occurred by operating multiple DHXs.

Journal Articles

Investigation on natural circulation behavior for decay heat removal in reactor vessel of sodium-cooled fast reactor under severe accident condition, 1; Effect of decay-heat conditions on natural circulation behavior under dipped-type DHX operation conditions

Tsuji, Mitsuyo; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu

Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 6 Pages, 2022/10

In sodium-cooled fast reactors (SFRs), decay heat removal after a core disruptive accident (CDA) is an important issue for the safety enhancement. Therefore, water experiments using a 1/10 scale experimental apparatus (PHEASANT) that simulates the reactor vessel of an SFR are conducted to investigate the natural circulation phenomena in the reactor vessel. In this study, experiments under the operation of the dipped-type DHX were conducted to investigate the effect of the heat generation ratio between the fuel debris on the core catcher in lower plenum and the reactor core remnant on the natural circulation behavior in the reactor vessel. The temperature distribution and the velocity distribution were measured under two heat generation conditions. Thus, the effect of the heat generation ratio between the fuel debris in the lower plenum and the reactor core remnant on the natural circulation behavior was quantitatively grasped under the dipped-type DHX operating conditions.

Journal Articles

Investigation on natural circulation for decay heat removal in reactor vessel of sodium-cooled fast reactor

Aizawa, Kosuke; Tsuji, Mitsuyo; Kobayashi, Jun; Kurihara, Akikazu; Miyake, Yasuhiro*; Nakane, Shigeru*; Ishida, Katsuji*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 10 Pages, 2022/04

In sodium-cooled fast reactors (SFRs), optimizing the design and operate decay heat removal systems (DHRSs) is important for safety enhancement against severe accidents that could lead to core melting. The natural circulation phenomena in a reactor vessel during operating a DHRS were clarified by conducting water experiments using a 1:10 scale experimental facility (PHEASANT) simulating the reactor vessel of loop-type SFRs. In this study, we investigated the natural circulation phenomena under conditions of operating the dipped-type DHX and RVACS using the results of temperature and particle image velocimetry (PIV) measurements, respectively. Furthermore, the effects of temperature fluctuation on the PIV measurement were quantitatively evaluated.

Journal Articles

Study on cooling process in a reactor vessel of sodium-cooled fast reactor under severe accident; Velocity measurement experiments simulating operation of decay heat removal systems

Tsuji, Mitsuyo; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu; Miyake, Yasuhiro*

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 5 Pages, 2020/08

The water experiments using a 1/10 scale experimental apparatus simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.

Journal Articles

Effects of temperature fluctuation on PIV measurement of natural circulation flow field

Tsuji, Mitsuyo; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu; Miyake, Yasuhiro*

Proceedings of 14th International Symposium on Advanced Science and Technology in Experimental Mechanics (14th ISEM'19) (USB Flash Drive), 4 Pages, 2019/11

The particle image velocimetry (PIV) was measured in scaled-model water experiments simulating a natural circulation flow field in a sodium-cooled fast reactor vessel. The temperature fluctuation in the natural circulation flow field causes the distribution of the refractive index. Thus, the temperature fluctuation affects the uncertainty of the velocity in the PIV measurement. In this study, the authors evaluated the effects of the temperature fluctuation on the PIV measurement in the natural circulation flow field.

Oral presentation

Study on cooling process of decay heat removal systems in a reactor vessel of sodium-cooled fast reactor under severe accident

Tsuji, Mitsuyo

no journal, , 

To elucidate the core cooling systems of a sodium-cooled fast reactor under severe accident, PIV experiments were carried out by using scaled water experimental facility simulating the condition of uniformly-accumulated debris on the core catcher, measured the natural convection flow field in starting up the submerged type DHX. The authors confirmed the specific flow patterns that low temperature fluid flowed down along the core wall and flowed into lower plenum, and the fluid was heated on the core catcher and elevated toward the center of reactor core, then some vortexes were formed by the mutual interaction due to the down-flow of low temperature fluid and up-flow of elevated temperature fluid. Furthermore, the equivalent flow field and the maximum flow velocity were observed compared with the existing knowledge.

Oral presentation

Study on cooling process of decay heat removal systems in a reactor vessel of sodium-cooled fast reactor by scaled water experiments, 2; PIV measurements of flow field in a reactor vessel simulating operation of dipped-type DHX

Tsuji, Mitsuyo; Ono, Ayako; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu; Miyake, Yasuhiro*

no journal, , 

Thermal-hydraulic phenomena driven by natural circulation in a reactor vessel was investigated by using scaled model water experiments simulating a reactor vessel in order to enforce of safety and optimize design and operation of decay heat removal systems under normal operation and severe accident conditions. This paper reports PIV measurement results of natural convection flow field in reactor vessel simulating the condition of uniformly-accumulated debris on the core catcher and operating the submerged type DHX.

Oral presentation

Study on cooling process of decay heat removal systems in a reactor vessel of sodium-cooled fast reactor by scaled water experiments, 3; Temperature measurement in a reactor vessel simulating operation of reactor vessel auxiliary cooling system

Aizawa, Kosuke; Tsuji, Mitsuyo; Kobayashi, Jun; Kurihara, Akikazu; Nakane, Shigeru*; Ishida, Katsuji*

no journal, , 

Thermal-hydraulic phenomena driven by natural circulation in a reactor vessel was investigated by using scaled model water experiments simulating a reactor vessel in order to enforce of safety and optimize design and operation of decay heat removal systems under normal operation and severe accident conditions. This paper shows temperature measurement results of natural convection flow field in reactor vessel simulating operation of reactor vessel auxiliary cooling system. In addition, it is shown that the affect of dispersal condition of molten fuel on temperature distribution in reactor vessel.

Oral presentation

Study on cooling process of decay heat removal systems in a reactor vessel of sodium-cooled fast reactor by scaled water experiments, 4; Effect of heat generation condition on natural convection flow field in reactor vessel

Tsuji, Mitsuyo; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu; Nakane, Shigeru*; Ishida, Katsuji*

no journal, , 

Thermal-hydraulic phenomena caused by the natural circulation in a reactor vessel were investigated using scaled model water experiments simulating the reactor vessel in order to enhance safety and optimize the design and operation of decay heat removal systems under normal operation and severe accident conditions. This report shows the measurement results of temperature and PIV of natural convection flow field in the reactor vessel simulating an operation of dipped type direct heat exchanger. In addition, it is shown that the effect of dispersal condition of molten fuel on natural convection flow field in the reactor vessel.

Oral presentation

Study on cooling process of decay heat removal systems in a reactor vessel of sodium-cooled fast reactor by scaled water experiments, 5; Transient behavior under operations of multiple decay heat removal systems

Aizawa, Kosuke; Tsuji, Mitsuyo; Kobayashi, Jun; Kurihara, Akikazu; Nakane, Shigeru*; Onuma, Hideyoshi*

no journal, , 

Thermal-hydraulic phenomena driven by natural circulation in a reactor vessel was investigated by using scaled model water experiments simulating a reactor vessel to enforce the safety and to optimize the design and operation of decay heat removal systems under normal operation and severe accident conditions. This paper shows the results of transient behavior under operations of dipped-type DHX and penetrated-type DHX to clarify the effect of operating multiple decay heat removal systems on natural circulation phenomena in reactor vessel. From the experimental results, it was confirmed that the maximum temperature temporarily increased compared to the maximum temperature in natural circulation steady state by operating the second DHX after the natural circulation steady state was maintained under the first DHX operation.

Oral presentation

Study on cooling process of decay heat removal systems in a reactor vessel of sodium-cooled fast reactor by scaled water experiments, 6; Thermal hydraulics behavior under operations of multiple decay heat removal systems

Tsuji, Mitsuyo; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu

no journal, , 

Thermal-hydraulic phenomena driven by natural circulation in a reactor vessel was investigated by using scaled model water experiments simulating a reactor vessel in order to enforce of safety and optimize design and operation of decay heat removal systems under normal operation and severe accident conditions. This report shows temperature and PIV measurement results of natural convection flow field in the reactor vessel under operations of dipped-type DHX and penetrated-type DHX. From the experimental result, it was conformed that the effect of the operation of the penetrated-type DHX on the natural convection behavior in the reactor vessel including the cooling of core and debris on core catcher.

Oral presentation

PLANDTL-2 experiment for evaluation of decay heat removal in sodium-cooled fast reactors; Effect of inter-subassembly gap flow under interruption of flow through fuel assemblies

Ezure, Toshiki; Akimoto, Yuta; Tsuji, Mitsuyo; Kurihara, Akikazu; Tanaka, Masaaki

no journal, , 

Sodium experiments using PLANDTL-2 test facility, which have multiple rows of simulated core, have been carried out to clarify the decay heat removal characteristics inside a reactor. In this study, decay heat removal experiments were carried out using a dipped-direct heat exchanger under no primary flow rate conditions and interrupting the flow through the assemblies by a plugging system of fuel assemblies. As the results, the influence of inter-subassembly gap flow on the cooling characteristics of simulated core was clarified.

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